Calculation of Reactivity-device Incremental Cross Sections Using MCNP Version 5

Abstract

This paper describes the logic that was implemented in a user-defined tally option TALLYX in MCNP-5 to calculate the scattering and the transport cross sections of reactivity devices. Incremental cross sections of the liquid-zone controllers and of the cobalt adjusters of a CANDU reactor were calculated with the user-defined TALLYX subroutine and the standard tally cards. The MCNP-based incremental cross sections of the reactivity devices were used in reactor-core simulations. The results of the analysis showed that the average simulation error for the liquid-zone controllers was –7.4% for the MCNP-based incremental cross sections and –6.2% for the DRAGON-based Side-Step method. The methodology for using TALLYX in MCNP-5 is now considered ready for general use.